In a pressurized water reactor (PWR), the heat generated by the nuclear reaction on the primary side is passed to reactor coolant. The reactor coolant is, in turn, passed into a hot leg of a steam generator (e.g., a Westinghouse Model 51 steam generator), moves through U-tubes (the “tube bundle”) in the steam generator (passing heat into water on the secondary side of steam generator), and passes out through a “cold” leg of the steam generator to return to the reactor to complete and perpetuate the loop.
The PWR steam generator typically is an upright cylindrical pressure vessel with hemispherical end sections, clad on internal surfaces with austenitic stainless steel. A transverse plate, a tubesheet, is disposed at a lower end of the cylindrical section, dividing the steam generator into a primary side, which is the lower hemispherical section below the tubesheet, and a secondary side above the tubesheet. Two 16 inch diameter manways permit access for inspection and maintenance during outage periods. A vertical wall bisects the primary side into an inlet section and an outlet section. The tubesheet also supports the U-tubes, noted above, at their lower ends. The tubesheet comprises thousands of holes (e.g., 11,252 holes) on the inlet section and the outlet section with one end of each U-tube inserted into a hole on the inlet section and the other end of the U-tube inserted into a hole at the outlet section (e.g., 5,626 tubes). Thus, each U-tube communicates, on one end, with the inlet section of the primary side, and the other end communicates with the outlet section of the primary side. The U-tubes are supported at various elevations above the tubesheet by support plates (e.g., seven tube support plates, roughly 1 inch thick and spaced about 39 inches apart). Tube support plates comprise openings for each U-tube, such support plates comprising, for example, an egg crate type configuration or a quatrefoil (cloverleaf shaped) configuration, with the support plates are themselves supported by tie rods and the steam generator cylindrical wrapper. Two small handholes are located above the tubesheet to provide limited access to the secondary side of the tubesheet.
The steam generator cylindrical wrapper, noted above, is disposed between the tube bundle and the steam generator shell and defines therebetween an annular “downcomer” chamber, terminating a predetermined distance above the tube sheet. A secondary fluid (“feedwater”) is introduced into an upper portion of the secondary side of the steam generator to travel down the annular chamber formed between the inner surface of the steam generator cylinder and the outside surface of the steam generator wrapper and to the tube sheet, radially inwardly along the tube sheet and upwardly among the U-tubes tubes inside the wrapper to circulate around the U-tubes, above the tubesheet, in a heat transfer relationship with the outside of the U-tubes. Owing to this heat transfer, a portion of the feedwater is converted to steam and the steam is then passed to steam turbines to generate power.
There are various mechanisms that are recognized to cause degradation of the safety of the barrier between the primary side and the secondary side and/or the efficiency of transfer of heat from the primary side to the secondary side, such as general corrosion, intergranular corrosion, pitting, and stress corrosion cracking (SCC), fretting, denting, etcetera.
Title 10 of the Code of Federal Regulations (10 CFR) establishes the fundamental regulatory requirements with respect to the integrity of the steam generator U-tubes, stating in the general design criteria (GDC) in Appendix A of 10 CFR Part 50 that the reactor coolant pressure boundary (RCPB) between the primary side and the secondary side shall be “designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture” (GDC 14), shall be “designed, fabricated, erected, and tested to the highest quality standards practical” (GDC 30), and shall be “designed to permit periodic inspection and testing of important areas and features to assess their structural and leaktight integrity” (GDC 32). Given the importance of steam generator tube integrity, PWR licensees have technical specifications governing the surveillance of steam generator U-tubes, but such technical specifications typically do not prescribe nondestructive test methods for inspecting tubes or specify where a particular methodology should be used, as is noted in Generic Letter (GL) 2004-01 issued by the U.S. Nuclear Regulatory Commission. As further noted therein, steam generator U-tubes are also subject to the quality assurance requirements of 10 CFR Part 50, Appendix B. Notwithstanding that the technical specifications (TS) do not specify nondestructive test methods or in what locations particular test methods must be employed, Criterion IX of 10 CFR Part 50, Appendix B, “Control of Special Processes,” requires, in part, that nondestructive testing be controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and other special requirements.
NRC Generic Letter 97-06 deals with steam generator U-tube integrity and failure mechanisms related to U-tube support degradation including, for example, corrosion-based degradation of steam generator tube eggcrate supports, discovered through secondary side visual inspections performed during an outage. NRC Generic Letter 97-05 also dealt with steam generator tube inspection techniques. Earlier, NRC Information Notice (IN) 79-27 highlighted the potential of steam generator U-tube ruptures as a result of foreign material in the secondary side of steam generators. Guidance on steam generator inspection is provided by Electric Power Research Institute (EPRI) guidelines and NEI 97-06, which is a generic self imposed industry guideline that lays out inspection, maintenance and repair criteria for steam generators.
To satisfy such requirements and guidelines, inspections were previously carried out by manually pushing video probes into the steam generator, which was and is disadvantageously radiation dose intensive (requires worker protection and limitation of exposure time), or by using inspection crawlers, which could inspect the annulus region, but could not inspect the annular region and the in-bundle high-flow region between the tubesheet and the first tube support.